• Economic and Environmental Geology
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• v.55 no.3
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• pp.219-229
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• 2022

Bentonite is being considered as a candidate for buffer material in geological disposal systems for high-level radioactive wastes. In this study, the effect of cement-bentonite interactions on bentonite alteration was investigated by reviewing the literature on studies of cement-bentonite interactions. The major bentonite alteration by hyperalkaline fluids produced by the interaction of cementitious materials with groundwater includes cation exchange, montmorillonite dissolution, secondary mineral precipitation, and illitization. When the hyperalkaline leachate from the reaction of the cementitious material with the groundwater comes into contact with bentonite, montmorillonite, the main component of bentonite, is dissolved and a small amount of secondary minerals such as zeolite, calcium silicate hydrate, and calcite is produced. When montmorillonite is continuously dissolved, the physicochemical properties of bentonite may change, which may ultimately causes changes in bentonite performance as a buffer material such as adsorption capacity, swelling capacity, and hydraulic conductivity. In addition, the bentonite alteration is affected by various factors such as temperature, reaction period, pressure, composition of pore water, bentonite constituent minerals, chemical composition of montmorillonite, and types of interlayer cations. This study can be used as basic information for the long-term stability verification study of the buffer material in the geological disposal system for high-level radioactive wastes.

• The Korean Journal of Nuclear Medicine Technology
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• v.14 no.2
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• pp.128-132
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• 2010

Purpose: Nuclear medicine general operation room is radioactive control room which is used for the handling of radioisotope(R.I). Radioactive contamination materials must be under control and separated from general trash. With this experiments, we want to actively suggest the guideline of controling and operating radioactive contamination materials by measuring contamination degree and analyzing the causes which is not realized so far. Materials and Methods: Materials are selected from Oct. 2009 to March. 2010. salines which are used for labelling radiophamaceuticals and generator cap, saline needle cap, $^{99m}Tc$-needle cap saline vial which is generated from $^{99}Mo$/$^{99m}Tc$ generator. After measuring each surface contamination degree by survey meter, mean value and standard deviation one were solved out. Results: In result, After measuring surface contamination degree, radioactivity of saline for labelling radiophamaceuticals showed $14429{\pm}26378$ cpm (p<0.05) and in measured generators, foreign imported things showed that generator cap : $9{\pm}21$ cpm, saline vial : $17{\pm}28$ cpm. saline needle cap : $35{\pm}66$ cpm, $^{99m}Tc$-needle cap : $9{\pm}21$ cpm, saline vial $13{\pm}28$ cpm. domestic things showed that generator cap : $22852{\pm}52545$ cpm, saline needle cap : $87367{\pm}109711$ cpm, $^{99m}Tc$-needle cap : $9008{\pm}10459$ cpm, saline vial : $186416{\pm}158196$ cpm (p<0.05). Conclusion: The saline which is used for labelling, exceeded 1/10 of maximum permissible range. this is generated from radiophamaceuticals dilution procedure. and In generators, radioactive value of foreign import things showed closely background value. but which of domestic thing showed that exceeded more than 1000 values 1/10 of maximum permissible range. the causes of that is domestic generator is contaminated in manufacturing procedure. So, to dispose radioactive contamination materials which is could betaken out of, the control and operation must be radical under controlled by radioactive measuring, recording and equipping of its own. if this is kept well, we can prevent surely that radioactive waste could be disposed like as general trash.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.343-356
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• 2016

The Korea Radioactive Waste Agency (KORAD) has developed a dual-purpose metal cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. The metal cask was designed in compliance with international and domestic technology standards, and safety was the most important consideration in developing the design. It was designed to maintain its integrity for 50 years in terms of major safety factors. The metal cask ensures the minimization of waste generated by maintenance activities during the storage period as well as the safe management of the waste. An activation evaluation of the main body, which includes internal and external components of metal casks whose design lifetime has expired, provides quantitative data on their radioactive inventory. The radioactive inventory of the main body and the components of the metal cask were calculated by applying the MCNP5 ORIGEN-2 evaluation system and by considering each component's chemical composition, neutron flux distribution, and reaction rate, as well as the duration of neutron irradiation during the storage period. The evaluation results revealed that 10 years after the end of the cask's design life, $^{60}Co$ had greater radioactivity than other nuclides among the metal materials. In the case of the neutron shield, nuclides that emit high-energy gamma rays such as $^{28}Al$ and $^{24}Na$ had greater radioactivity immediately after the design lifetime. However, their radioactivity level became negligible after six months due to their short half-life. The surface exposure dose rates of the canister and the main body of the metal cask from which the spent nuclear fuel had been removed with expiration of the design lifetime were determined to be at very low levels, and the radiation exposure doses to which radiation workers were subjected during the decommissioning process appeared to be at insignificant levels. The evaluations of this study strongly suggest that the nuclide inventory of a spent nuclear fuel metal cask can be utilized as basic data when decommissioning of a metal cask is planned, for example, for the development of a decommissioning plan, the determination of a decommissioning method, the estimation of radiation exposure to workers engaged in decommissioning operations, the management/reuse of radioactive wastes, etc.

• The Journal of Engineering Geology
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• v.26 no.1
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• pp.79-85
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• 2016

A concrete silo plays an important role in subsurface low- and intermediate-level waste facilities (LILW) by limiting the release of radionuclides from the silo geosphere. However, due to several physical and chemical processes the performance of the concrete structure decreases over time and consequently the concrete loses its effectiveness as a barrier against groundwater inflow and the release of radionuclides. Although a number of processes are responsible for degradation of the silo concrete, it is determined that the main cause is corrosion of the reinforcing steel. Therefore, the time it takes for the silo concrete to fail is calculated based on two factors: the initiation time of corrosion, defined as the time it takes for chloride ions to penetrate through the concrete cover, and the propagation time of corrosion. This paper aims to estimate the time taken for concrete to fail in a LILW disposal facility. Based on the United States Department of Energy (DOE) approach, which indicates that concrete fails completely once 50% of the volume of the reinforcing steel corrodes, the corrosion propagation time is calculated to be 640 years, which is the time it takes for corrosion to penetrate 0.640 cm into the reinforcing steel. In addition to the corrosion propagation time, a diffusion equation is used to calculate the initiation time of corrosion, yielding a time of 1284 years, which post-dates the closure time of the LILW disposal facility if we also consider the 640 years of corrosion propagation. The electrochemical conditions of the passive rebar surface were modified using an acceleration method. This is a useful approach because it can reduce the test time significantly by accelerating the transport of chlorides. Using instrumental analysis, the physicochemical properties of corrosion products were determined, thereby confirming that corrosion occurred, although we did not observe significant cracks in, or expansion of, the concrete. These results are consistent with those of Smartet al., 2006 who reported that corrosion products are easily compressed, meaning that cracks cannot be discerned by eye. Therefore, it is worth noting that rebar corrosion does not strongly influence the hydraulic conductivity of the concrete.

• Journal of Radiation Protection and Research
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• v.24 no.2
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• pp.101-108
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• 1999

The influence of the relative humidity, the temperature and the velocity of supply air on evaporation rate has been studied with non-boiling forced evaporation system in order to treat very low level radioactive liquid wastes produced from the decontamination and decommissioning activities. Experimental data on the evaporation rate have been obtained with the divers variables and experimental equation of air velocity was also obtained by the correlation of those data. The decontamination factor of this system was also obtained by the experimental data from a simulated liquid waste containing Cs-137 radio isotope ; $DF=10^4$. Since the commercial system will be operated for the treatment of the very low level radioactive liquid waste produced from decontamination & decommissioning of TRIGA Mark-II&III research reactor, the environmental assessment has been conducted to improve the operational safety. Exposure dose rate for an individual member of general public was assessed, and it showed that it was very lower than individual dose limits. The release of radioactivity of radioisotope material (Cs-137) to the environment was assessed, and result showed that it was $4.637{\times}10^{-14}\;{\mu}Ci/cc$.

• Journal of the Korean Recycled Construction Resources Institute
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• v.9 no.4
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• pp.451-459
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• 2021

The purpose of this study is to evaluate the performance of a solidifying agent for recycling the fine powder separated from the nuclear power plant decommissioned concrete as a solidifying agent(SA) for radioactive waste. In order to evaluate the performance of the solidifying agent, a powder simulating the fine powder of waste concrete separated from the dismantled concrete of a nuclear power plant was produced, and the main variables were the type of binder and the replacement ratio of zeolite. The solidifying agent was evaluated for fluidity performance, compressive strength, and leaching resistance to non-radioactive cesium. The compressive strength of SA increased as the zeolite replacement ratio increased, and the SA containing 5% or more of zeolite showed a compressive strength that was 1.4 to 1.7 times higher than the acceptance criteria. The cesium leaching index of all specimens was 6 or higher, satisfying the acceptance criteria, and the leaching index of SA was 1.47~1.63 times higher than that of OPC. In particular, the average leaching index after 28 days of the 5% zeolite-substituted solidifying agent was 9.15, which was improved by about 6.4% compared to OPC, and it was confirmed that the zeolite was effective in improving the leaching resistance to cesium ions by showing stable performance over the entire period.

• Journal of the Korean Society of Radiology
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• v.13 no.2
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• pp.297-303
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• 2019

The nuclear medicine department of a domestic medical institution uses $^{99m}TcI$, a radionuclide, from $^{99}Mo/^{99m}TcI$ Generator, to inject radioactive drugs into patients. Among the expired generators, imported from foreign countries, the medical institution implements its own disposal. Each medical institution shall satisfy the permitted in-house disposal concentration of radioactive wastes. The guidelines for self-disposal presented in Korea suggested that self-disposal can be performed 80 days after the generator is used. The purpose of these guidelines is to analyze them by comparing them with the data measured directly with the generator and to study if they are feasible. As a result, the generator with a capacity of 1,000 mCi has the longest half-life, and when tested with a high-radiation Mo(molybdenum) column, the number of days that are below the permitted concentration of body disposal with radioactive waste was 72 days and 71 days that were derived from direct column measurement. The results of the direct study confirmed that the guidelines for in-house disposal in Korea were reasonable, as there were 8 to 9 days of storage compared to the number of in-house disposal days provided in the guidelines.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.2
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• pp.95-102
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• 2013

There was a study for biological characteristics, except for physico-chemical and mineralogical properties, on the natural bentonite that is considered as a buffer material for the high-level radioactive waste disposal site. A bentonite slurry that was prepared from a local 'Gyeongju bentonite' in Korea was incubated in a serum bottle with nutrient media over 1 week and its stepwise change was observed with time. From the activated bentonite in the nutrient media, we can find a certain change of both solid and liquid phases. Some dark and fine sulfides began to be generated from dissolved sulfate solution, and 4 species of sulfate-reducing bacteria (SRB) were identified as living cells in samples that were periodically taken and incubated. These results show that sulfate-reducing (or metal-reducing) bacteria are adhering and existing in the powder of bentonite, suggesting that there may be a potential occurrence of longterm biogeochemical effects in and around the bentonite buffer in underground anoxic environmental conditions.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.407-417
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• 2004

This study is to verify radiation detection method by using $\alpha$-spectroscopy and ${\gamma}$-spectroscopy for concretes and components which will be generated during the decommissioning of the uranium conversion plant. Components and inside walls of the building were contaminated with natural uranium materials. Some parts of the stainless steel pipes and concretes of the walls were sampled and analyzed their alpha and gamma activities respectively. Alpha and gamma activities are well matched each other in the range of high activity region to 0.01 Bq/g and gamma activities are over estimated comparing alpha activities corresponded in below 0.005 Bq/g region for the natural uranium of AUC sample. The $^{238}U$ originated from natural products of conversion process could be distinguished by measuring $^{214}Pb$ or $^{214}Bi$ and $^{234}Th$ or $^{234m}Pa$. Uranium contaminations mainly are in the wall surface of the plant. Decontamination process of generating wastes which can be reached tp background level gamma activities measured by gamma spectroscopy can also be used to conservative assessment data.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.21-36
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• 2020

With respect to spent nuclear fuels, disposal containers and bentonite buffer blocks in deep geological disposal systems are the primary engineered barrier elements that are required to isolate radioactive toxicity for a long period of time and delay the leakage of radio nuclides such that they do not affect human and natural environments. Therefore, the thermal stability of the bentonite buffer and structural integrity of the disposal container are essential factors for maintaining the safety of a deep geological disposal system. The most important requirement in the design of such a system involves ensuring that the temperature of the buffer does not exceed 100℃ because of the decay heat emitted from high-level wastes loaded in the disposal container. In addition, the disposal containers should maintain structural integrity under loads, such as hydraulic pressure, at an underground depth of 500 m and swelling pressure of the bentonite buffer. In this study, we analyzed the thermal stability and structural integrity in a deep geological disposal environment of the improved deep geological disposal systems for domestic light-water and heavy-water reactor types of spent nuclear fuels, which were considered to be subject to direct disposal. The results of the thermal stability and structural integrity assessments indicated that the improved disposal systems for each type of spent nuclear fuel satisfied the temperature limit requirement (< 100℃) of the disposal system, and the disposal containers were observed to maintain their integrity with a safety ratio of 2.0 or higher in the environment of deep disposal.